ORAU TEAM

Dose Reconstruction

Project for NIOSH

 

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 34

Document Title:

Site Profile for Allied Chemical Corporation Plant

Document Number: ORAUT-TKBS-0044

Revision: 01 

Effective Date: 10/01/2007

Type of Document: TBD

Supersedes: Revision 00 

Subject Expert(s): Shelby L. Gubin, Robert Hysong, Cindy W. Bloom, and Joseph S. Guido

Site Expert(s): N/A

Approval:

Signature on File

Approval Date:

09/20/2007

Bernard M. Olsen, Document Owner

Approval: Signature on File Approval Date: 09/20/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 09/20/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 09/27/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 10/01/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 2 of 34

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

02/01/2006 00 New document to establish the technical basis for the development 

of a radiation exposure matrix for Allied Chemical Corporation Plant. 

First approved issue. Training required: As determined by the Task 

Manager. Initiated by Cindy W. Bloom.

10/01/2007 01 Approved Revision 01 initiated to add text in Section 2.0 and remove

information regarding nonuranium radionuclides in Sections 3.0 and 

5.0 to incorporate SEC evaluation report recommendation to not 

reconstruct nonuranium internal dose. Moved Figure 2-1. 

Incorporates formal internal and NIOSH review comments. Added 

reference for employee statement regarding X-rays. Constitutes a 

total rewrite of the document. The recently designated residual 

period is identified. Section 5.0, Estimating Exposures From 

Residual Radioactivity, expanded to address dose reconstruction 

methodologies for exposures received during the residual period. 

Added Attributions and Annotations section and additional 

references. Training required: As determined by the Task Manager. 

Initiated by Bernard M. Olsen.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 3 of 34

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

5

1.0 Purpose and Scope..................................................................................................................7

2.0 Site Description and Operational History.................................................................................. 8

2.1 Site Description ............................................................................................................8

2.2 Process Description....................................................................................................11

2.3 Source Term...............................................................................................................16

2.4 Safety.........................................................................................................................18

2.4.1 Visual Observation of Contamination .............................................................. 18

2.4.2 Air Activity Measurement of Contamination..................................................... 19

2.4.3 Surface Contamination Measurements ........................................................... 19

2.4.4 Decontamination............................................................................................. 19

2.5 Incidents.....................................................................................................................19

2.6 Physical Examinations – X-Rays ................................................................................19

2.7 Summary Operational Period Assumptions, Workdays, Work 

Hours, Work Categories .............................................................................................20

2.8 Residual Contamination..............................................................................................20

3.0 Estimating Internal Exposure ................................................................................................. 20

3.1 Uranium......................................................................................................................20

3.1.1 Uranium Urinalysis Data ................................................................................. 21

3.1.2 Uranium In Vivo Data...................................................................................... 21

3.2 Notations on Bioassay Records..................................................................................23

3.3 Occupational Internal Dose Reconstruction Assumptions and 

Summary....................................................................................................................23

4.0 Estimating External Exposure ................................................................................................ 23

4.1 Occupationally Required Medical X-Ray.....................................................................24

4.2 Miscellaneous Information about External Dose .........................................................25

4.3 Occupational External Dose Reconstruction Assumptions and 

Summary....................................................................................................................25

5.0 Estimation of Exposure from Residual Radioactivity............................................................... 25

6.0 Attributions and Annotations .................................................................................................. 27

References ........................................................................................................................................29

ATTACHMENT A, APPLICATION OF RESIDUAL DOSE FACTORS..................................................32
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 4 of 34

LIST OF TABLES

TABLE TITLE

2-1 Radionuclide activity ratios in relation to uranium based on reported activities in 

ACCP concentrates, wastes, air, and effluents.......................................................................

PAGE

17

3-1 Natural uranium’s specific activity and isotopic fractions ........................................................ 20

3-2 Mean whole body 137Cs from fallout in the United States........................................................ 22

4-1 External exposure summary................................................................................................... 25

5-1 Source term activity ratios relative to uranium (based on the largest reported ratio 

in relation to uranium in Table 2-1) ......................................................................................... 26

5-2 Adjustment factors to account for depletion of pre 1977 source term during the 

residual contamination period ................................................................................................ 26

5-3 Residual radiation exposure summary ................................................................................... 27

LIST OF FIGURES

FIGURE TITLE

2-1 Plot plan of ACCP in Metropolis, Illinois .................................................................................

PAGE

10

2-2 Process steps for conversion from ore concentrates (yellowcake) to UF6............................... 11
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 5 of 34

ACRONYMS AND ABBREVIATIONS

ACCP Allied Chemical Corporation Plant

AEC U.S. Atomic Energy Commission

AP anterior-posterior

AWE atomic weapons employer

Ci curie

cm centimeter

DOE U.S. Department of Energy

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft feet

hr hour

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

KOH potassium hydroxide

L liter

m meter

MeV megaelectron-volt, 1 million electron-volts

mL milliliter

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

ORAU Oak Ridge Associated Universities

pCi picocurie

RMC Radiation Management Corporation

SEC Special Exposure Cohort

U.S.C. United States Code

yr year

μCi microcurie

μg microgram
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 6 of 34

§ section or sections
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 7 of 34

1.0 PURPOSE AND SCOPE

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer [AWE] facility” or a “Department of Energy [DOE] facility” as defined in the Energy 

Employees Occupational Illness Compensation Program Act of 2000 [EEOICPA; 42 U.S.C. § 7384I(5) 

and (12)]. EEOICPA, as amended, provides for employees who worked at an AWE facility during the 

contract period and/or during the residual contamination period.

Employment at an AWE facility is categorized as either (1) during the contract period (i.e., when the 

AWE was processing or producing material that emitted radiation and was used in the production of 

an atomic weapon), or (2) during the residual contamination period (i.e., periods that NIOSH has 

determined there is the potential for significant residual contamination outside of the period in which 

weapons-related production occurred). For contract period employment, all occupationally derived 

radiation exposures at covered facilities must be included in dose reconstructions. NIOSH does not 

consider the following exposures to be occupationally derived:

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

For residual contamination period employment, only the radiation exposures defined in 42 U.S.C. 

§ 7384n(c)(4) [i.e., radiation doses received from DOE/U.S. Atomic Energy Commission (AEC)-

related work] must be included in dose reconstructions. Radiation dose received from DOE/AECrelated work includes: (1) radiation from radon consistent with NIOSH policies for including such 

radiation in the contract period; and (2) medical screening X-rays, but not diagnostic X-rays for the 

treatment of work-related injuries. It should be noted that: (1) under subparagraph A of 42 U.S.C. § 

7384n(c)(4), radiation associated with the Naval Propulsion Program is specifically excluded from the 

employee’s radiation dose; and, (2) under subparagraph B of this section, radiation from a source not 

covered by subparagraph A that cannot be reliably distinguished from radiation that is covered by 

subparagraph A is considered part of the employee’s radiation dose. This site profile covers only 

exposures resulting from nuclear weapons-related work. Exposures resulting from non-weaponsrelated work, if applicable, will be covered elsewhere.

A decision has been made that internal dose from nonuranium radionuclides cannot be reconstructed 

with sufficient accuracy for employees of the Allied Chemical Corporation Plant (ACCP) during the 

period 1959 through 1976, so only internal dose from uranium is considered in Section 3 of this site 

profile. The Secretary of Health and Human Services (HHS) has designated a single class of 

employees from ACCP for inclusion in the Special Exposure Cohort (SEC) (Leavitt 2007). The class 

of employees for which NIOSH cannot estimate radiation doses with sufficient accuracy includes all 

AWE employees who were monitored or should have been monitored for exposure to ionizing 

radiation while working at the ACCP in Metropolis, Illinois, for a number of workdays aggregating at 

least 250 workdays from January 1, 1959, through December 31, 1976, or in combination with 

workdays within the parameters established for one or more other classes of employees in the SEC. 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 8 of 34

Section 2.0 of this document describes the Allied site and its history including information about the 

radiological processes and source terms as well as the radiological controls and monitoring practices. 

Sections 3.0 and 4.0 discuss internal and external dose assessment, respectively. Section 5.0 

provides information for assigning dose during the residual radiation period beginning in 1977, after 

the cessation of AEC related operations.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.0.

2.0

The information that follows applies to the period of AEC operations at ACCP in Metropolis, Illinois, 

from January 1, 1959, to December 31, 1976, which involved AEC-contracted conversion of uranium 

ore concentrates to uranium hexafluoride. Opening ceremonies for the Allied Chemical’s fluorine 

plant occurred in October 1958 (Sloop 1978). The original license issue date was December 17, 1958 

(NRC 2003, p. 8-9). Four claims (Claim Numbers redacted) include bioassay results in December 

1958, and one of these from December 15, 1958, has a positive result that indicates that uranium 

exposures might have occurred in late 1958, before the DOE Office of Worker Advocacy-established 

covered period. The plant was closed temporarily on June 30, 1964, and AEC (1966) reported that it 

might reopen in 1966. A review of available worker dosimetry records indicated that some workers 

might have been laid off or transferred around this period, but that other workers were still on the site. 

Documentation reviewed indicates that weapons related residual contamination exists outside the 

listed operational period (NIOSH 2006b). Residual contamination from prior weapons-related 

activities is indistinguishable from contamination produced during subsequent operations. This facility 

is still operational, and the residual contamination period has been established as being from January 

1, 1977 through the present (NIOSH 2006b).

SITE DESCRIPTION AND OPERATIONAL HISTORY

Detailed information about buildings and processes is provided in Sections 2.1, 2.2, and 2.3 and is 

based on descriptions written after the period of operations covered by this document. Individuals 

who worked at ACCP indicated that there was little variation in the processes over time. Because 

individual dosimetry results are available for the vast majority of workers at ACCP, the reconstructed 

doses will be based on the specific individual dosimetry.

The ACCP radiological source term started with the receipt of uranium ore concentrates. Some longlived uranium progeny were included in the concentrates (e.g., 230Th and 226Ra). Uranium chemical 

forms included oxides, fluorides, and hexafluorides, which meant that exposures could have been to 

uranium solubility types F, M, and/or S. 

The ACCP was also known as General Chemical Division, Allied Signal Metropolis Plant, and was 

later purchased by Honeywell.

2.1 SITE DESCRIPTION

“On February 4, 1957, the Allied Chemical and Dye Corp. announced selection of Metropolis Ill., as 

the site of its plant to process 5,000 tons of U3O8 a year under contract with the [Atomic Energy] 

Commission” (AEC 1957). The official plant startup date appears to be January 1, 1959, but there are 

indications that uranium and uranium workers might have been on the site in late 1958. One 

individual reported participating in opening ceremonies at Allied Chemical's fluorine plant at 

Metropolis, Illinois in October 1958 (Sloop 1978), which indicates that uranium exposure could have 

occurred as early as October 1958. 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 9 of 34

Allied Chemical operated its UF6 production facility using a dry conversion process (versus a wet 

solvent extraction process) to supply UF6 feed for the Paducah Gaseous Diffusion Plant through June 

30, 1964, under an AEC contract (Perkins 1982). In addition, Allied Chemical was sampling uranium 

concentrates for other entities as of 1982.

On June 30, 1964, the plant was temporarily closed. AEC (1966) reported that the plant “may be 

reopened in 1966 for uranium hexafluoride production.” Perkins (1982) noted that ACCP resumed 

operations in February 1968, but the available dosimetry records indicate that radiological exposures 

were still occurring at the plant between June 30, 1964, and February 1, 1968. Some of the claimant 

interviews and external dosimetry records show that a number of workers were on the site during the 

shutdown, and hiring was occurring in 1967, although these records also indicate that a number of 

workers were laid off in July 1964. The available records do not indicate what activities might have 

been ongoing in the plant during the shutdown period. However, based on U.S. Department of Labor 

job descriptions associated with one claim (Claim Number redacted), some processing operations 

might have continued at least on a sporadic basis. 

The ACCP U.S. Nuclear Regulatory Commission (NRC) license (NRC 2003, p. 8-2) states: 

The Allied Signal Metropolis Plant is located on approximately 1,000 acres of land in 

Massac County at the southern tip of Illinois, along the north bank of the Ohio River. 

The site perimeter is formed by US Highway 45 to the north, the Ohio River to the 

south, an industrial coal blending plant to the west and privately owned, developed land 

to the east. Plant operations are conducted in a single fenced-in, restricted area 

covering 59 acres in the north-central portion of the site.

The plant is now owned by Honeywell and is still providing UF6 to gaseous diffusion plants for use in 

the commercial sector. The 1982 building locations are shown in Figure 2-1. 

The following description of the buildings is excerpted from ACCP licensing documentation (NRC 

2003, pp. 8-8 to 8-9). The list was originally numbered 8.5.1 through 8.5.8, but has been simplified to 

1 through 8 below.

Most of the uranium processing equipment is housed in a six-story structure termed the 

Feed Materials Building where essentially all of the steps in the UF6 manufacturing 

process are conducted. Other areas and buildings in which operations are conducted 

involving the handling or processing of significant quantities of source material include 

the following:

1. A Sampling Plant which receives samples or concentrates for uranium assay 

and moisture content.

2. The Sodium Removal (pretreatment) and Uranium Recovery Facilities which 

are housed in buildings where high sodium content ore concentrates are 

treated to remove sodium impurity, and where materials are reprocessed to 

recover additional uranium.

3. The KOH [potassium hydroxide] muds washing facility which removes fluorides 

and KOH from the potassium diuranate muds generated in the fluorination 

scrubber system. The washed potassium diuranate is then processed through 

Sodium Removal. The wash liquors are neutralized at EPF [environmental 

protection facility, this facility probably did not exist during the early years]. 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 10 of 34

Figure 2-1. Plot plan of ACCP Metropolis, Illinois (Perkins 1982, p. 3).

4. The Calcining Facility which dries the incoming feed material and recovered 

uranium as the first step in ore preparation.

5. The Pond Mud Calciner Drier where hard/wet ore concentrates and KOH Muds 

are processed prior to packaging for blending with additional ore concentrates 

at the Feed Materials Building for conversion to UF6. 

6. The Laboratory Building which houses facilities for conducting process control, 

product, and radiological control analyses.

7. The Cylinder Wash Building where UF6 product cylinders are periodically 

washed and hydrostatically tested prior to reuse.

8. Outdoor pads for storage of drums of ore concentrates and other uranium 

bearing materials, as well as UF6 product cylinders.

Additional plant facilities which are involved directly in the UF6 manufacturing process 

but do not involve the handling of any significant quantities of source material include a 

fluorine manufacturing building, a fluoride waste treatment facility with four large CaF2

settling tanks, a powerhouse, a reductor off-gas incinerator, and two small uranium 

settling ponds to collect any uranium spills.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 11 of 34

Allied Chemical was also involved in the manufacture of 30,000 pounds per week of liquid fluorine, 

1,200 tons per year of sulfur hexafluoride, 2,500 pounds per week of antimony pentafluoride, and 

10,000 pounds per week of iodine pentafluoride (Perkins 1982). These materials were not 

radioactive.

2.2 PROCESS DESCRIPTION

The conversion process from ore concentrates (yellowcake) to UF6 is described below, and a process 

flow diagram is presented in Figure 2-2. As of 1982, the processing capacity was 14,000 tons of 

uranium per year and the plant had processed more than 100,000 tons of uranium from domestic and 

foreign uranium concentrates (Perkins 1982).

As part of an evaluation of environmental control technologies for nuclear fuel [uranium hexafluoride 

(UF6)] conversion facilities, Perkins (1982) described the conversion of yellowcake to uranium 

hexafluoride at ACCP. Although some changes in the process might have occurred since the first 

days of operation, it is likely that these changes would have been related to ore sampling techniques, 

ventilation and filtration controls, and waste disposal practices, and that the general process was 

similar throughout the plant’s history. Process descriptions from Perkins (1982) follow.

Figure 2-2. Process steps for conversion from ore concentrates 

(yellowcake) to UF6 (NRC 2003, p. 9-18).

Introduction

The Allied facility uses the fluoride volatility process to produce UF6 from uranium 

concentrates. The facility does not accept uranium slurries.

When the 55-gallon-drummed concentrate is received it is weighed, sampled, and 

analyzed for moisture content. After feed preparation, mechanically sized particles 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 12 of 34

(U3O8) are reacted with hot cracked ammonia to form uranium tetrafluoride (UF4) or 

green salt. Next, the green salt is combined at high temperature in a fluidized bed with 

gaseous fluorine to form the gas UF6. Finally, the UF6 off-gas is further purified by 

condensation and fractional distillation.

Each of these steps, with the resulting waste and effluent streams, is discussed in 

detail in the following sections.

Receiving and Sampling

Uranium concentrates are received at Allied Chemical Corp. Plant in 55-gallon drums. 

The drums are stored outside on specially constructed pads designed to allow 

rainwater to drain to centralized collection sumps, which is then pumped to uranium 

spill control ponds. 

Drums are weighed, and a falling stream method is used to obtain a representative 

sample. The airborne dusts generated in the sampling process are removed through 

the use of two baghouses in series, which discharge through a stack. Empty drums 

are air cleaned and the discharge runs through two baghouses before being 

discharged to a stack.

Concentrates that are found to contain a high percentage of sodium are pretreated in 

the sodium-removal facility.

Sodium Removal (Pretreatment)

Because sodium forms a compound that causes caking and sintering in the fluorination 

fluid beds, any incoming uranium concentrates containing sodium and all the uranium 

concentrates produced by the uranium recovery sections of the UF6 facility must first 

have the sodium or potassium removed before they can be processed. These 

concentrates are treated by reaction with a solution of ammonium sulfate in four 

counter current vessels to make (NH4)2U2O7 for further processing. The chemical 

reaction is

Na2U2O7 + (NH4)2SO4 → (NH4)2U2O7 + Na2SO4

The liquid effluent contains the excess (NH4)2SO4, Na2SO4, K2SO4, and uranium and 

small quantities of 226Ra and 230Th, selenium, molybdenum, and other trace impurities. 

The effluent is sent to the two uranium spill control ponds. Uranium contaminated 

storm water is also discharged to these ponds. The sludges are allowed to settle, and 

the decantate is discharged to the main discharge outfall for the plant.... When the 

minimum freeboard is reached (approximately 2 ft) on a pond, the pond is dredged and 

the sludge is sent to the pond sludge calciner.... Pond liners are inspected and 

repaired at this time. In addition, all pond liners are underlain by a gravel layer that 

allows any seepage to drain to a leak detection sump....

The (NH4)2U2O7 (ADU) [ammonium diuranate] is removed from solution in the last 

reaction vessel and is sent to the main feed preparation section of the plant. 
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Feed Preparation

In the feed preparation circuit, sodium-free uranium concentrate, either from the 

pretreatment section or from the sampling section, is first fed into a calciner where 

water and ammonia are removed. The off-gases containing ammonium, sulfur dioxide, 

uranium, and contaminant particles pass through two baghouses in series. 

Particulates recovered by blowback of the baghouses are collected and sent back into 

the blending feed stream.... 

The calcined material is blended and then agglomerated, dried, crushed, and sized 

before being sent to the reduction circuit. Wet off-gas streams join the off-gases from 

the calciner, whereas dried off-gases pass through two baghouses in series before 

discharge.... Material collected by the dry stream baghouse is fed into the blending 

feed stream. 

Reduction

In the reduction circuit, Allied operates two trains in parallel, both having the same 

basic design. The U3O8 from the feed preparation circuit is fed into the reductor vessel 

where hot cracked ammonia (N2 and H2) and additional nitrogen mix with the feed to 

form a fluidized bed. The U3O8 is reduced to UO2, which is withdrawn from the bottom 

of the bed. Off-gases from the reactor include H2, N2, H2S, AsH3, vaporized S, SeH2, 

and particulates composed of UO2, unreacted U3O8 and reduced compounds of 

impurities originally in the concentrate.... 

The off-gases pass through two porous metal filters in series, a sulfur condenser, and 

incinerator.... 

Particulates collected by cleaning the filters are sent to the uranium recovery section.

Off-gases from the oxide vacuum cleaner are sent through cyclone and two baghouses 

in series before discharge.... Residues from these collection devices are sent to the 

uranium recovery section.

Hydrofluorination

Again in the hydrofluorination circuit, two trains in parallel are in use. The UO2 from 

reduction is fed into two reactor vessels in series, which use vaporized hydrofluoric 

acid and N2 to fluidize the UO2 and permit HF to react with the material to form UF4. 

The off-gases contain the excess HF needed to give good conversion to UF4 (10% 

excess or more), volatile SiF4, BF3, and some of the molybdenum and vanadium (which 

were present as impurities in the concentrate) as volatile fluorides and oxyfluorides, 

and any remaining sulfur as H2S4. These gases first pass through two sets of porous 

carbon filters per set, where the unreacted UO2 particles, entrained UF4 particles, and 

any other particulate matter are removed. The gases then pass through two venturi 

water scrubbers where the H2O in the off-gases condenses, and finally through a 

venturi KOH scrubber and a packed tower using KOH scrubbing liquid.

The residue from the clean carbon filters is sent to uranium recovery. Liquid coming 

from the venturi water scrubbers contains hydrofluoric acid and is sent to the acid 

neutralization treatment plant, where lime is added to precipitate the fluorine as CaF2. 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 14 of 34

The uranium content of this stream is reported to average less than 5 ppm [parts per 

million]. Silicon, boron, and some molybdenum, and vanadium, which were originally 

in the concentrate, are expected to be in the sludge. The liquid from the KOH venturi 

and packed tower is sent to the liquid treatment systems.

The vacuum cleaning off-gases from the cleaning activities necessary in the 

hydrofluorination section are routed through a cyclone and two baghouses before 

discharge.... Residues from the dust collectors are sent to the uranium recovery 

section.

Fluorination

The green salt, UF4, produced in the hydrofluorination process is fed into one of two 

fluid bed reactors used in parallel. Again, Allied uses two trains at the works. The bed 

material is CaF2, and unreacted UF4. The vessels are operated at temperatures of

approximately 1000° F, and good temperature control is required. Fluorine, F2, from an 

on-site fluorine production facility is introduced into the reactor vessels to convert UF4

into gaseous UF6. 

The off-gasses, including uranium UF6, F2, HF, unreacted UF4, and other particulates, 

and volatilized impurities ... are first cooled before passing through two sets of sintered 

nickel filters, each containing two filters. The material recovered as the filters are 

cleaned is stored, and then sent to the uranium recovery section. 

The stream exiting from the filters has the UF6 removed by condensation in a set of 

three cold traps used in series. Any uncondensed UF6, F2, HF and other volatiles 

passing from the cold wraps are removed from the gas stream by a KOH spray tower, 

a KOH pack tower, and finally a KOH coke box.

Carbon dioxide in the spent scrubber liquor reacts with the uranium and potassium as 

follows.

2(UO3)

4 - + 6CO2 + 6OH- + 2K+ → K2U2O7 + 6CO3

2- +3O2 + 3H2O 

The K2U2O7 is settled from the KOH solution and sent to the pretreatment facility. 

Decantate from the uranium settling section is sent to the KOH regeneration system as 

indicated previously.... The KOH is then sent back to the scrubbing system. In 

addition to CaF2 compounds of uranium, vanadium, and molybdenum, trace quantities 

of compound of silica, carbon sulfur, and other trace contaminants are expected to be 

in the sludge.... 

The condensed material in the three cold traps in series is melted and drained to the 

still feed tanks. Entrained hydrogen fluoride is vaporized during melting and passes to 

the off-gas system. 

From the still feed tanks, the liquid is fed to a low-temperature boiler bubble cap 

column in which the impurities are volatilized and exit from the top of the column.... 

The liquid uranium hexafluoride passes from the low boiler column into the bubble cap, 

high boiler column. In this column, the UF6 is volatilized and exits from the top of the 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 15 of 34

column. Non-volatilized impurities are removed from the bottom of the column and 

stored. 

The gaseous UF6 from the column is condensed into two cold traps operated in series. 

The UF6 is transferred to the shipping cylinders by melting the UF6 and allowing the 

material to drain into the cylinders.

Because impurities build up on the recycled CaF2 bed material used in the fluorination 

reactors, part of the bed material must be withdrawn periodically. This “ash” [also 

referred to as spar or bed material] along with the fluorination-cleaned filter residue, is 

drummed and stored for a minimum of 6 months to permit the uranium daughters 234Th 

and 234Pa to decay. The ash is then sent to the uranium recovery section. A total 

quantity of about 0.1 ton of ash per ton of uranium processed is produced, and it 

contains approximately 1.8% of the total uranium processed and most of the uranium 

daughters originally present in the concentrate.

The dusts produced by withdrawing bed material from the reactors are passed through 

a cyclone and two baghouses in series before discharge.... The ash vacuum cleaner 

systems also discharge from this stack. Residue from cleaning these particulate 

removal devices is treated similarly to the spent bed material.... 

The feed preparation, reduction, hydrofluorination, and fluorination circuits are all 

located in the feed material building. This building has a complete air changeout

approximately once every 5 minutes.... 

Washdown of the feed materials building is pumped from collection sumps into the 

uranium recovery ponds.

The transfer lock to fluorination discharges through two baghouses.... Again, dust 

collection residue is sent to uranium recovery. 

Uranium Recovery

There are several sources of uranium-containing wastes produced by the facility, and it 

is desirable to recover the uranium from these wastes. Thus, uranium-containing dried 

sludges, spent bed material, filter residue, and various scraps and dust are sent to the 

uranium recovery circuit. 

Off-gases from storage before processing vent through the 3-1 stack. The main 

contaminants in the stack off-gases should be radon and radon daughters, if the 

radium originally in the concentrate feed goes into the ash (as was assumed in the 

previous section).

The material from storage is first dumped and ground. Dusts from this operation are 

recovered in a baghouse.... 

The finely ground material is leached using a sodium carbonate solution to solubilize 

the uranium as the tricarbonate complex.... 
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After leaching, the pregnant solution is filtered. Lime is used to aid in precipitation of 

any fluoride, and the solution is again filtered. Sludge from the filters is dried. Offgases from the dryer pass through the dumping and grinding off–gas baghouse.

The dried sludge probably contains all nondecayed uranium daughters, some residual 

uranium, and some fluoride compounds. This sludge is drummed and sent to a 

licensed radioactive waste disposal facility. Approximately 1500 tons of this waste, 

containing a maximum of 46.9 Ci is produced yearly. 

The uranium is precipitated from the solution using NaOH. The uranium is filtered and 

washed and then sent to the sodium removal section.

Cylinder Wash

Cylinders are returned to Allied from the enrichment facilities. These cylinders contain 

residual UF6 and the daughters of uranium that have “grown in” while the cylinders 

containing UF6 were in storage. The daughters in general are found plated out on the 

cylinder walls.

After any residual UF6 is removed from the cylinder, the cylinder is washed, using a 

solution of Na2CO3 to remove any impurities from the walls and to solubilize the 

uranium. The wash solution is filtered to remove the unleached solids, and the 

pregnant solution is pumped to join the pregnant solution in the uranium treatment 

section.

The solid residue from the filters contains daughter products of uranium, principally 

234Th and 234Pa, and is stored on-site in drums until disposal in a licensed waste 

disposal facility.

Sludge Dryer

As previously described, the process generates several sludges that contain uranium. 

Before being sent for uranium recovery, these sludges are dried in a calciner. Offgases containing SO2, HF and particulates from the calcining operation pass through a 

baghouse and a water spray tower before discharge.... 

The spent scrubber water, which contains uranium, fluoride compounds, and some 

sulfur compounds, is sent to the uranium recovery ponds.

The calcined sludges are drummed and stored until they can be processed in the 

uranium recovery section.

2.3 SOURCE TERM

The plant produced UF6 from uranium ore concentrates initially for gaseous diffusion facilities and 

later also for commercial fuel fabrication facilities. The uranium conversion process at the plant has 

remained essentially unchanged since its inception. In 1982, Perkins (1982) reported that the 

concentrates initially contained 70% to 75% uranium by mass, and these concentrates included 

impurities that were dependent on the mines and mills from which the yellowcake came. 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 17 of 34

Uranium chemical forms at ACCP included all solubility types (i.e., F, M and S). The initial design 

processing capacity was 5,000 tons of uranium in a year, and by 1982 the throughput capacity had 

been increased to 14,000 tons of uranium in a year in what was essentially a continuous chemical 

process. As of 1982, more than 100,000 tons of uranium had been processed (Perkins 1982). 

The initial milling of the uranium before arrival at ACCP resulted in disequilibrium of uranium progeny 

in the concentrate. Exposure to the nonuranium radionuclides occurred at the front end of the 

processing, beginning with receipt. Increased concentrations of other radionuclides in relation to 

uranium were found in the fluorination bed ash and other solid waste streams, such as the sludge. 

Environmental air measurements in the late 1970s showed ratios of 226Ra and 230Th in relation to 

uranium that were larger than the concentrate ratios but lower than complete equilibrium ratios.

Radon concentrations would probably have been elevated during the initial opening of the drums of 

concentrate, in the feed area, and in work areas where radium would have concentrated. No 

measurements of radon concentrations have been found at this time. 

Perkins (1982) summarized uranium, thorium, and 226Ra activities in ACCP effluents and wastes from 

the 1970s. These values, listed in Table 2-1, were used to estimate ratios of the activity of other 

radionuclides to uranium in the effluent. The reported concentrate ratio was based on an average of 

yellowcake feed ratios from 33 mills (Perkins 1982, Table II-8). The equilibrium activity ratio of 230Th 

to uranium in uranium ore would be about 0.4886; the milling of the ore usually reduces this ratio by 

almost a factor of 100 by removing the 230Th from the concentrate. It should be noted that 232

Th 

comprises a portion of the radioactivity in the uranium concentrates, and although 232

Th was not 

specifically reported for most effluents at ACCP, the water effluents for 1975 and 1976 showed annual 

average 232

Th:U-nat and 230Th:U-nat activity ratios of 0.076 and 0.017, respectively. 

Table 2-1. Radionuclide activity ratios in relation to uranium based on reported activities in ACCP 

concentrates, wastes, air, and effluents.

Source

Average Largest ratio

Th-nat:U-nat Th-230:U-nat Ra-226:U-nat Th-nat:U-nat Th-230:U-nat Ra-226:U-nat

Concentrate

(yellowcake feed 

averaged over 33 

mills)

ND 5.20E-03 ND ND ND ND

Uranium recovery solid 

waste (dried sludge)

3.58E+00 1.90E+01 1.11E+00 ND ND ND

Solid nonsludge waste ND 5.27E-03 2.75E-04 ND ND ND

Outdoor air ND 6.70E-02 1.97E-03 ND 1.67E-01 8.46E-03

Air effluent ND 1.70E-02 2.70E-04 ND 4.75E-02 3.38E-04

Water effluent ND 3.99E-03 2.04E-02 ND 1.58E-02 5.79E-02

ND: No data provided in Perkins (1982)

Processing at ACCP results in further separation of uranium and its progeny, and this can result in 

increased activity ratios of nonuranium radionuclides in some areas in the plant. The largest 

disequilibrium ratios during the operational period are associated with the solid waste (dried sludge) 

from the uranium recovery facility. As stated above, these materials were barreled and shipped offsite 

as radioactive waste. AEC and DOE inventory reports use a designation of JSA to identify ACCP. 

DOE reports that the Rocky Flats Plant shipped 71 kg (160 pounds) of depleted recycled uranium to 

ACCP in 1975, which is a very small fraction of the annual inventory (which might have been as high 

as 10 to 28 million pounds) processed at ACCP (RFETS 2000, Appendix A, p. 27). A second mention 

of the same shipment indicates that the 71 kg were shipped to Allied Chemical Corporation in Idaho 

Falls, Idaho (RFETS 2000, Appendix B, p. 2). Another reference (Author unknown ca. 2000) referred

to a shipment of 22,060 kg from Rocky Flats Plant to ACCP between 1955 and 1964, but this appears 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 18 of 34

to be an error due to a mix-up between the next designation in the report, LAB, with JSA. It is 

assumed that contaminants in recycled uranium were not significant contributors to ACCP workers’ 

radiation dose.

ACCP is licensed by the NRC to have sealed Cs-137 sources up to 100 millicuries each. These are 

used as level control devices and are unlikely to contribute to internal dose.

2.4 SAFETY

Specific radiological safety measures implemented in the beginning of ACCP operations are not 

available but, as of 1958, radiological control programs in general had become more sophisticated. 

Dosimetry is available from the earliest days of covered operations for all employees working in the 

controlled area. The following information is based primarily on reference material dated after 1972, 

and it should be noted that the criteria to prompt radiological control activities could have differed in 

the early years.

Engineered confinement was and is the primary source of contamination control used at ACCP. 

Drums, process vessels, lines, etc., are used to confine the materials. When unconfined uranium is 

handled, procedures call for ventilation controls to reduce exposures (Author unknown 1985, p. 1). 

An exception was requested from the NRC requirement to post and label individual radioactive 

material areas and containers, which was replaced by a requirement to train workers (NRC 2003, p. 

11-1). Signs state that any area or container in the plant may contain radioactive materials.

Protective clothing and shoes or shoe covers were provided to employees and visitors to ACCP to 

ensure that personnel were not inadvertently contaminated with uranium compounds as of 1982 

(Perkins 1982), and probably from the earliest days of operations. This did not apply to administrative 

areas where contamination would not be expected.

Current ACCP procedures, discussed below, include four contamination measurement and control 

parameters:

• Visual observation of contamination

• Air concentration measurements

• Surface contamination measurements 

• Decontamination procedures 

2.4.1

Allied Chemical noted that natural uranium compounds have a low specific activity, characterized by a 

high ratio of mass to activity. The ACCP uranium compounds are highly colored, which provides an 

immediate method for detecting gross surface and air contamination. For instance, yellowcake was 

yellow and UF6 releases appeared as white smoke due to the decomposition products, UO2F2 and 

HF. Procedures from 1985 stated that if employees detected small spills, they were responsible for 

cleanup (Author unknown 1985). Larger leaks or spills that resulted in visible accumulation of 

uranium on equipment or on the floor required immediate decontamination, activation of the respirator 

lights, and possibly shutdown of the unit. The foremen were to prepare a health physics incident 

report and notify the health physics shift technician to initiate special air sampling and investigation as 

appropriate. Health physics personnel made a daily inspection of uranium operating areas to ensure 

proper decontamination and contamination control. These procedures were consistent with 

contemporary industrial health practices for uranium facilities (DOE 1988). 

Visual Observation of Contamination
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 19 of 34

2.4.2

As of 1985, stationary breathing zone samples were collected daily from 57 stations in the uranium 

hexafluoride building, 10 stations in the sampling plant, 2 stations in the drum dumper, 3 stations in 

sodium removal, and 2 stations in the uranium recovery unit. Summaries of some of these 

measurements can be found in the available references. As of 1985, respirators were required when 

the average gross alpha air concentration on a floor exceeded 4E-11 μCi/cm3 or when any four air 

samples on the floor exceeded that concentration. In addition, as of 1985, any single air sample 

result that exceeded 1E-10 μCi/cm3 required investigation and preparation of a health physics incident 

report (Author unknown 1985, p. 3).

Air Activity Measurement of Contamination

Although some air concentration data are available for the 1970s and later, these data are not used to 

estimate worker intakes because of the abundance of bioassay data.

2.4.3

Surface contamination measurements in uranium and nonuranium areas are made routinely. 

Removable contamination is determined by taking a wipe or smear of 100 cm2 on a surface. Each 

smear is counted for alpha radioactivity and reported in dpm/100 cm2 removable alpha contamination. 

The plant lunchroom and offices or control rooms are surveyed weekly, as are the locker rooms and 

personnel exit areas. Uranium processing areas such as the Feed Materials Building, Sodium 

Removal, Uranium-Recovery Building, and Sampling Plant are surveyed monthly. Other nonuranium 

processing areas are surveyed quarterly (Author unknown 1985; Author Unknown 2002a).

Surface Contamination Measurements

As of 1985, decontamination is to be initiated when removable alpha contamination exceeds 

5,000 dpm/100 cm2 in the uranium processing areas. All other areas are to be decontaminated when 

removable alpha contamination exceeds 200 dpm/100 cm2 (Author unknown 1985). 

One summary of Health Physics data for the last half of 1991 (Wilkins 1992) noted that 1.7% of the 

2,002 weekly smears exceeded the weekly limit of 200 dpm/100 cm2 and that the highest result was 

923 dpm/100 cm2 in the lunchroom on a table. One of the 570 monthly smears and none of the 163 

quarterly smears exceeded the limit. 

2.4.4

Allied Chemical has detailed procedures for decontaminating spills or leaks found in the plant. These 

procedures generally include roping off the area, sweeping or vacuuming the contaminated material, 

and then washing the contaminated surface. During these procedures protective clothing is to be 

used and monitoring is to be performed (Author unknown 1985; Author unknown 2002b).

Decontamination

2.5 INCIDENTS

The claims include some information about incidents, but details are few and there are no dates or 

references to particular incident reports. Occasional leaking valves or inadequate packing resulted in 

material releases. 

2.6 PHYSICAL EXAMINATIONS – X-RAYS

No documented ACCP requirement for occupationally required medical X-ray examinations has been 

found. More than half of the claimant interviews indicate that chest X-ray examinations were routine 

and a number of the interviews indicate the frequency was annual. The examinations were offered 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 20 of 34

annually, but were optional by 1980, and would have been done at Massac Memorial Hospital or 

perhaps a local doctor’s office according to one employee, who began work at ACCP in the 1970s 

(Bloom 2005). 

2.7 SUMMARY OPERATIONAL PERIOD ASSUMPTIONS, WORKDAYS, WORK HOURS, 

WORK CATEGORIES 

Because bioassay and film badge data were used to estimate internal and external exposures during 

the AEC uranium operations, an estimate of the number of workdays or work hours per year is not 

necessary. 

No attempt has been made to sort workers into exposure categories. Individual dosimetry appears to 

be available for most workers. 

2.8 RESIDUAL CONTAMINATION 

Uranium concentrate processing continued at ACCP after 1976, and differentiation between residual 

contamination from the AWE period and contamination due to later and/or ongoing operations is not 

feasible. 

3.0

As stated in Section 1.0 above, a decision has been made that internal dose from nonuranium 

radionuclides cannot be reconstructed with sufficient accuracy for employees of ACCP during the 

AEC operations period 1959 through 1976. The primary source of internal radiation exposure at 

ACCP was uranium dust produced from the processing of uranium concentrates to produce uranium 

hexafluoride. It is assumed that the uranium was of natural enrichment before 1977, although there is 

some indication that the ACCP source term included a small amount of depleted uranium. Although 

the uranium ore concentrate contained uranium progeny of dosimetric interest as well as thorium 

(assumed to be 232Th and 228Th), only uranium dose is estimated for ACCP claims for the period

January 1, 1959 through December 31, 1976.

ESTIMATING INTERNAL EXPOSURE

3.1 URANIUM

During the processing of uranium concentrates at ACCP, workers might have been exposed to a 

variety of uranium chemical forms that encompass all lung absorption types: F, M, and S. Although 

some process steps might have had more or less exposure to a given lung absorption type, it is not 

clear how well separated these areas were. In addition, workers might have worked in multiple 

uranium process areas. The selection of absorption type should depend on the organ of interest. The 

specific activity of the natural uranium and the isotopic mass and activity fractions are listed in 

Table 3-1 [1]. For organ dose estimates, the uranium activity can be assumed to be from 234U [2]. A 

conversion factor is provided to adjust results reported as μg of 235U to total uranium activity in 

picocuries.

Table 3-1. Natural uranium’s specific activity and isotopic fractions.

Isotope

Activity 

fraction

Mass 

fraction

Activity ratio 

to U-235

Mass ratio 

to U-235

U-234 0.4886 5.37E-05 21.4 0.00745

U-235 0.0228 7.20E-03 1 1

U-238 0.4886 9.93E-01 21.4 138

Useful factors

U-natural pCi/μg 0.683
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 21 of 34

U (pCi): U-235 (μg) 94.8

3.1.1

Individual uranium urinalysis began as early as December 1958 and was done in house. It appears 

that there are bioassay data for all workers. The worker records show that bioassays were typically 

collected at hire and periodically thereafter. Allied Chemical notes that fluorimetry was used for 

analyses until 2002, when kinetic phosphorescence analysis (KPA) began (Rogers 2006). Urinalysis 

data for 1989 to 1992 are missing, and ACCP reported that it was assumed that the missing records 

might have been water-damaged beyond recovery.

Uranium Urinalysis Data

In the early years, uranium urinalysis results reported in microcuries per milliliter (μCi/mL) are 

misstated; the units are actually micrograms per liter (μg/L). The recording limit of 10 μg/L is found in 

the early records and is consistent with the default detection threshold for uranium urinalysis assumed 

to be 10 μg/L based on a reported sensitivity of 5 to 10 μg/L for uranium fluorimetry urinalysis in the 

early years (Wilson 1958). Beginning in late 1974, a recording limit of 5 μg/L appears to have been in 

use. Allied Chemical reported that its minimum detection level (MDL) was 2 μg/L in the early 1980s, 

but no records of MDL are available before that. In 1985, ACCP noted that routine urinalysis samples 

were to be left after 24 to 96 hours of absence from the workplace, but it is unknown when this 

procedure was implemented. In addition, as of 1985, ACCP employees were encouraged to leave 

special urine samples before their next work shift in response to known or suspected airborne 

exposures (NRC 1993, p. C-14). In later years, it was noted that a result equal to 30 μg/L was cause 

for investigation.

On the computer printouts, results are sometimes noted in the comments section rather than the 

bioassay measurement section. 

3.1.2

Lung counts are available for many workers beginning in late 1978, which is after the covered 

operational period. Lung counts appear to have been performed on an annual or biannual frequency 

until the mid-1980s. They appear to be less frequent in later years, but this could indicate that only 

some workers were required to have annual lung counts in later years.

Uranium In Vivo Data

ACCP claims include a letter that reports use of a uranium lung-counting system from Radiation 

Management Corporation (RMC) and analysis of the spectra using a software package called 

RMCWBC, which was based on the Oak Ridge National Laboratory program ALPHA-M (Rogers 

2006). The counting system includes an RMC-designed shadow shield and four detectors, which 

appear to consist of thin NaI(Tl), and thicker CsI(Na) crystals. The diameter is reported as 5 feet, but 

is probably 5 inches. Before analysis of the spectra, the background is automatically subtracted. A 

REMCAL standard man equivalent phantom was used to calibrate the system. Radionuclides include 

235U, 234Th, 137Cs, and 40K.

Specific ACCP criteria are given for judging the data and can be used for considering the relative 

validity of a result (Rogers 2006):

• Fit should be less than 3.0 and usually ranges between 0.5 and 1.5. An ideal fit would be 0.

• The shift value should range between +2.0 and -2.0; values outside the range of ±3.0 are 

carefully reviewed before they are accepted. MCA zero level and/or gain adjustments are

warranted when the shift value exceeds ±2.0.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 22 of 34

• The gain should ideally be 1.0 and a gain of 0.90 to 1.10 is considered acceptable. 

Adjustment to the gain and/or zero might be needed if the value is outside this range.

The analysis program reportedly prints a table showing all radionuclides found and their estimated 

activities and relative 2-standard-deviation counting error in terms of a percent error. Allied Chemical 

reports that results of 235U are given in μg and results for other radionuclides are given in nanocuries; 

however, some records show 234Th results in μg. Results with errors greater than 100% (results less 

than 2 times their counting standard deviation) are considered questionable. The reported 235U 

minimum detection level was noted as 63 μg, which was equated to 32% of the maximum permissible 

lung burden. The maximum permissible lung burden was noted as 194.3 μg of 235U or summed as 

26,974 μg of total uranium (Author unknown 1986, p. 11). In later years, it was noted that a result 

equal to 50% of the maximum permissible lung burden was cause for investigation.

Unless there is a prominent low-energy peak or the detector ratios vary by more than 0.025, the 

detected uranium is assumed to be in the lung.

In general, the urinalysis results should be used to estimate intakes and the lung counts can be used 

as supporting data. The lung counts do not directly measure total uranium and, although height and 

weight are recorded and could have been used to estimate chest wall thickness, there is no indication 

that chest wall thicknesses were measured directly.

To convert 235U mass to total uranium activity, it is assumed that the uranium is of natural enrichment, 

and so the 235U mass in μg is multiplied by 2.16 pCi/μg (the specific activity of 235U) and divided by 

0.0228 (the activity fraction of 235U), a total factor of 94.8, to obtain the total uranium activity in 

picocuries [1].

To convert 234Th to total uranium activity, it is assumed that the 234Th is in equilibrium with both 238

U 

and 234

U, and so the mass of 234Th in micrograms is multiplied by 2.32E+10 pCi/μg and divided by 

0.4886 to obtain the total uranium activity in picocuries [3]. If the 234Th is given in nCi, it can be 

converted to picocuries of uranium by dividing by 4.886E-4. Note that with a 24.1-day half-life, 234Th 

might not be in equilibrium with its uranium parent depending on when and where an exposure 

occurred.

The 40K and 137Cs reported in lung counts are not likely to be associated with workplace exposures; 

40K is ignored and any positive 137Cs results are evaluated against the background levels due to 

fallout. Cesium-137 values associated with fallout are shown in Table 3-2.

A handwritten note (Author unknown 1996) indicates that other whole-body counting systems might 

have been in use as of 1996, but there were no differences in reported sensitivity. 

Table 3-2. Mean whole body 137Cs from fallout in the United States.a

Year Body burden (nCi) Year Body burden (nCi)

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6.0 1975 1.1
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 23 of 34

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. Mean whole body Cs-137 activity in the United States due to fallout. Adjusted 

from becquerels to picocuries by multiplying NCRP Report 94(NCRP 1987, Table 

B.5) results by 27 picocuries per becquerel.

3.2 NOTATIONS ON BIOASSAY RECORDS

Some of the bioassay records include notations and abbreviations. These include:

• Laid-off

• Rehired

• Trnfd [transferred]

• MTO [Morristown Office, corporate headquarters in New Jersey]

• MPLB [maximum permissible lung burden]

• Sample type R [assumed to mean routine]

• Sample type S [assumed to mean special]

3.3 OCCUPATIONAL INTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

The AEC operational exposure period was from January 1, 1959, to December 31, 1976. As stated in 

Section 1.0 above, a decision has been made that internal dose from nonuranium radionuclides 

cannot be reconstructed with sufficient accuracy for employees of ACCP during this period, so only 

internal dose from uranium is considered in this section.

In general, the urinalysis results should be used to estimate uranium intakes and the lung counts can 

be used as supporting data. The uranium photofluorimetry urinalysis reporting limit in the early years 

at ACCP was 10 μg/L [4]. The less-than results, when available in the individual dosimetry records, 

can be assumed to be uranium urinalysis detection thresholds. Uranium can be absorption type F, M, 

or S. For dose calculations, uranium intakes can be assumed to be 234U [2]. 

To estimate uranium intakes from the lung counts, the intake of 235

U is determined and then adjusted 

to total uranium intake activity using the conversion in Table 3-1. The lung detection threshold is 

63 μg for 235

U and 5,970 pCi for total uranium.

Based on the current information, all workers were monitored for uranium intakes, so at this time no 

assumptions have been provided for unmonitored workers or unmonitored periods. This section will 

be developed further, if necessary. 

4.0

External dosimetry results are available for ACCP workers beginning on January 1, 1959, and include 

results for gamma, beta, and neutron exposures. Allied Chemical reports that the data before 1970 

are handwritten and the records do not include the company or method of measurement. Beginning 

in 1969, R.S. Landauer Jr. & Company provided film badge dosimetry services and measured gamma 

and beta radiation. In about April 1976, Eberline Instrument Company provided thermoluminescent 

dosimeter services. Neither Landauer nor Eberline reported doses below 10 mrem.

ESTIMATING EXTERNAL EXPOSURE
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 24 of 34

Film badge detection thresholds in the early days were assumed to be as high as 50 mrem for gamma 

and 60 to 80 mrem for beta. A detection threshold of 40 mrem has been recommended for this 

Project (ORAUT 2006). 

The majority of photons from natural uranium have energies in the range of 30 to 250 keV, so this 

analysis assumed workers were exposed to photon energies in the 30- to 250-keV range, which is 

consistent with NIOSH (2006c). Nonpenetrating dose from natural uranium consists primarily of 

electrons with energies above 15 keV. Because the uranium is mixed with material that can produce 

neutrons from an alpha particle collision, the neutron dose is also considered. For consistent 

presentation, exposure or dose is reported as: 

• Penetrating, assumed to be associated with photons of energies 30 keV or greater

• Nonpenetrating, assumed to be associated with photons of energies less than 30 keV or with 

electrons

• Neutron, assumed to be associated with neutrons of energies of 0.1 to 2 MeV

The earliest available ACCP external dosimetry results are dated January 1, 1959. Badge exchange 

frequencies at ACCP were typically monthly in the available records. In the 1980s, hourly employees 

exchanged badges monthly and salaried employees changed badges quarterly.

Although some records have results recorded as neutron doses, in later years no records of neutron 

doses are available. Most of the results appear to be zeros, and it is not clear if these represent 

actual dosimeter results or were entered because there was no result. The Bridgeport Brass site 

profile (ORAUT 2005a) notes that the minimum detectable neutron dose reported in the early 1960s 

was 60 mrem. 

A reasonable neutron dose estimation which is favorable to the claimant is based on a 500-pound 

drum that contains 75% by mass of natural uranium. It is assumed that the alpha particles collide with 

a fluorine target. Using the method described in (ORAUT 2005b), the neutron dose is 3.28E-04 

rem/hr, calculated at 1 foot from a drum. A worker who is 1 foot from such a drum for 2,000 hr/yr 

would be exposed to a dose rate of 6.56E-01 rem/yr. This document considers this estimated neutron 

dose rate to be an upper bound (maximum estimate) of the ACCP neutron dose rate. The mode 

neutron dose rate is estimated as 5.47E-03 rem/yr, which is the result of applying an occupancy of 3 

hours/week over 50 weeks to the estimated dose rate at 3 feet from a drum [5]. The minimum 

neutron dose rate is estimated as 0 rem/yr. These estimated minimum, mode and maximum neutron 

dose rates are to be used only for reconstructing doses during the listed operational period, and 

applied in IREP as a triangular distribution. During the residual period, the weapons-related source 

materials (uranium, fluorine targets) needed to produce neutrons are not presumed to be present in 

amounts significant enough to warrant the estimation of neutrons. 

4.1 OCCUPATIONALLY REQUIRED MEDICAL X-RAY

No definitive information was found that indicated the type or frequency of occupationally required 

medical X-ray examinations, although the claimant interviews support an assumption of an annual 

chest X-ray examination. The type and projection of X-ray examination should be based on current 

Oak Ridge Associated Universities (ORAU) Team guidance, which at this time is a PA radiographic 

chest examination. Organ doses can be obtained from the current revision of ORAUT-OTIB-0006, 

Dose Reconstruction from Occupationally Related Diagnostic X-Ray Procedures (ORAUT 2005c).
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 25 of 34

4.2 MISCELLANEOUS INFORMATION ABOUT EXTERNAL DOSE

This section includes external dose information that could be of value for specific dose 

reconstructions. This analysis did not consider such information generically because of its limited 

applicability or because of limited details.

The fluorination bed ash contains short-lived progeny, which results in elevated beta and gamma 

radiation fields in the vicinity of the ash. Workers who handled this material or washed drums or 

cylinders could receive extremity doses that are significantly higher than those measured by 

dosimeters worn on the trunk of the body. 

4.3 OCCUPATIONAL EXTERNAL DOSE RECONSTRUCTION ASSUMPTIONS AND 

SUMMARY

Table 4-1 summarizes assumptions about occupational external doses during the covered period 

January 1, 1959, to December 31, 1976, at ACCP.

 Table 4-1. External exposure summary.

Exposure 

mode Exposure type Annual exposure

Uranium 

processing

Penetrating

(photons, 30–

250 keV, AP, 

acute)

Based on individual 

dosimetry

Non-penetrating

(electrons >15 

keV, acute)

Based on individual 

dosimetry 

Neutron

0.1 to 2 MeV, 

triangular

6.56E-01 rem max;

5.47E-03 rem mode;

0.00 rem min.

PA chest 

radiographic 

examination

Initial plus one 

exam per year

See ORAUT-OTIB0006 (ORAUT 

2005b) 

5.0

ACCP continued to operate after the covered AWE period ended on December 31, 1976. NIOSH has 

determined the residual radiation period to be from January 1, 1977 to present (NIOSH 2006b). As 

stated in Section 2.8, differentiation between residual contamination from the AWE period and 

contamination due to later and/or ongoing ACCP operations is not feasible. The following information 

provides a method for estimating exposures during the residual radiation period due to uranium and 

nonuranium residual contamination. 

ESTIMATION OF EXPOSURE FROM RESIDUAL RADIOACTIVITY

Internal exposures due to intakes of uranium after 1976 can be estimated using personnel monitoring 

data which was collected during this period (in vivo and in vitro data, as applicable). Table 5-1, below 

can be used to account for intakes of nonuranium radionuclides during the residual radiation period. 

The values in this table are based on the highest reported ratio in the outdoor air and air effluent 

pathways reported in Table 2-1. The use of these pathways during the residual period is appropriate 

since they are representative of the isotopic distribution of the resuspended materials to which 

workers present during the residual period would be exposed [6]. Since no values are reported for 

Th-232, the water effluent value (reported in section 2.3 were used). 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 26 of 34

Table 5-1. Source term activity ratios 

relative to uranium (based on the 

largest reported ratio in relation to 

uranium in Table 2-1).

U-natural 1

Th-232 0.076

Th-230 0.167

Th-228 0.076

Ra-226 0.00846

Pb-210 0.00846

Po-210 0.00846

In order to account for the fact that only a portion of the internal exposure during the residual 

contamination period is from resuspended residual contamination versus exposure due to continued 

site operations, calculated intakes are adjusted by a weighting factor to account for the continued 

depletion of the pre-1977 source term during the residual period. 

Decrease in particulate resuspension with time has been well documented in experimental studies in 

outdoor environments (Sehmel 1980, Till 1983). Measured resuspension factor “half-lives” in the 

range of 35 days to years have been reported (Sehmel 1980). Models for this effect have been 

proposed in the form of a constant (steady state component) and a second component with an 

exponential term. For example, Linsley reported an expression (Linsley 1978): 

K(t in days) = [10-6 exp(-0.01t) + 10-9]m-1 with the 10-6 factor being replaced by 10-5 

for periods of “regular disturbance by vehicular or pedestrian traffic.”

Fewer data are available on the variation of resuspension factors with time in indoor environments. 

However, Healy recommends a decay constant value of 0.1 d-1 which represents the effects of source 

depletion with time (Healy 1971). While no experimental studies were identified for indoor facilities, 

an exponential decrease in resuspension is expected to occur due to conservation of mass and the 

depletion of easily suspended contaminants.

Current project guidance (ORAUT-OTIB-0004) establishes a source term depletion factor of 1% of the 

surface activity per day. Use of this 1% depletion factor is favorable to claimants, based on the 

depletion behavior reported above, however, to account for the observed steady-state resuspension 

condition, this factor is held constant after 1979. The factors to account for depletion of the pre-1977 

source term during the residual period are presented in Table 5-2.

Table 5-2, Adjustment factors to 

account for depletion of pre 1977 

source term during the residual 

contamination period.

Year Factor

1977 1

1978 0.03

1979 - present 0.0007

Examples of the application of the Table 5-1 activity ratios and Table 5-2 depletion adjustment factors 

to estimate intake rates are presented in Attachment A.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 27 of 34

To estimate external exposure to residual activity, external dosimetry results that are available after 

1976 are used to provide an upper bound of the dose. Since external dose is from both resuspended 

and fixed contamination, no adjustment to external dose due to depletion of resuspended material is 

necessary.

To summarize, the information in Table 5-3 (supplemented with the detailed guidance in Sections 3, 

4, and 5) is used to account for dose during the residual exposure period.

Table 5-3. Residual radiation exposure summary.

Source Exposure information

Period of 

applicability Annual exposure

Uranium

U-234

Internal (alpha, 

chronic)

Beginning 

January 1, 1977

Inhalation intake based on 

individually calculated results 

adjusted by factors in Table 5-2.

Nonuranium

Th-232

Th-230

Th-228

Ra-226

Pb-210

Po-210

Internal (alpha, 

chronic)

Beginning 

January 1, 1977

Inhalation intake for listed 

nonuranium radionuclides based 

on the assigned uranium intake 

rate and the nonuranium ratio 

values of Table 5-1.

Uranium 

contamination

External

Penetrating (photons, 

30–250 keV, AP, 

acute)

Beginning 

January 1, 1977

Use reported dosimetry results

Uranium 

contamination

External

Nonpenetrating

(electrons >15 keV, 

acute)

Beginning 

January 1, 1977

Use reported dosimetry results

6.0 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database.

[1] Olsen, Bernard M. MJW Corporation. Senior Health Physicist. July 2007.

The data for natural uranium are as presented in the bioassay application, Integrated Modules 

for Bioassay Analysis (IMBA). 

[2] Olsen, Bernard M. MJW Corporation. Senior Health Physicist. July 2007.

This assumption results in estimated organ doses which are favorable to the claimant. The 

assumption that all uranium available for intake is 234U will result in slightly higher estimates of 

organ doses than if natural uranium is assumed. For a dose reconstruction employing best 

estimate methodology, natural uranium can be assumed.

[3] Olsen, Bernard M. MJW Corporation. Senior Health Physicist. July 2007.

The activity fraction of 234U and of 238U (0.4882) in natural uranium are as presented in the 

bioassay application, Integrated Modules for Bioassay Analysis (IMBA). The specific activity 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 28 of 34

of 234 Th (2.32 E+10 pCi/μg) and half lives of radionuclides are listed in any standard handbook 

of radiation protection quantities, e.g. Shleien, B., 1992, The Health Physics and Radiological 

Health Handbook, revised edition, Scinta, Inc., Silver Spring, MD.

[4] Olsen, Bernard M. MJW Corporation. Senior Health Physicist. July 2007.

The increased sensitivity of the urinalysis methods used at ACCP over time is evident from the 

individual bioassay monitoring records of ACCP claimants. Results less than the level of 

detection for urinalysis results, in units of micrograms per liter (μg/L), are reported as < 10 μg/L 

through approximately mid 1974; < 5 μg/L from then through approximately late 1978; and < 2 

μg/L after late1978.

[5] Olsen, Bernard M. MJW Corporation. Senior Health Physicist. August 2007.

Surveys of UF6 storage cylinders more enriched than those at Allied Chemical, and studies of 

work practices in moving those cylinders, have been made at the Portsmouth Gaseous 

Diffusion Plant, as described in ORAUT-2006d. It was estimated in ORAUT-2006d that a 

worker would take on average 3 hours/week, for 50 weeks in a year, to transport these 

cylinders, and that the average neutron dose rate to the worker would be that at approximately 

1 meter from the cylinder surface. The upper bound neutron dose rate developed in Section 

4.0 is 3.28E-04 rem/hr at 1 foot from the cylinder. At 3 feet from the cylinder, the neutron dose 

rate would be estimated at 3.28E-04 rem/hr [(1 ft) 2 / (3 ft) 2 ], or 3.64E-05 rem/hr. The 

average neutron dose rate, applying the 150 hr/year occupancy methodology of ORAUT2006d (3 hr/wk for 50 wk/yr), is 5.47E-03 rem/yr.

[6] Guido, Joseph S. MJW Corporation. Senior Health Physicist. July 2007. 

The airborne effluent and ambient air data from Perkins 1982 was collected between 1976 and 

1979 and as such are representative of the isotopic distribution at the start of the residual 

period.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 29 of 34

REFERENCES

AEC (U. S. Atomic Energy Commission), 1957, "Production of Special Nuclear Materials," Major 

Activities in the Atomic Energy Programs, January–June 1957, Washington, D. C., July, p. 9. 

[SRDB Ref ID: 9511, p. 13]

AEC (U. S. Atomic Energy Commission), 1966, "Private Nuclear Industry Growth," Annual Report to 

Congress of the Atomic Energy Commission for 1965, Washington, D.C., January, p. 37. 

[SRDB Ref ID: 9511, p. 17].

Author unknown, 1985, Policy and Procedure for Contamination Control, August 2. [SRDB Ref ID: 

12376, p. 27]

Author unknown, 1986, Policy and Procedure for Bioassay Sampling, January 27. [SRDB Ref ID: 

12376, p. 2]

Author unknown, 1996, Procedure for Bioassay Sampling, August 19. [SRDB Ref ID: 12378, p. 2]

Author unknown, ca 2000, page of summary information about Allied Chemical Corporation Plant. 

[SRDB Ref ID: 9511, p. 27]

Author unknown, ca. 2002a, Allied Chemical Corporation Plant procedures, Section 3.0: Surface 

Contamination Measurements. [SRDB Ref ID: 12378, pages 38-39] 

Author unknown, ca. 2002ba, Allied Chemical Corporation Plant procedures, Section 4.0: 

Decontamination. [SRDB Ref ID: 12378, p. 39] 

Bloom, C., 2005, phone conversation with former employee (ACCP), November 2.

DOE (U.S. Department of Energy), June 1988, Health Physics Manual of Good Practices for Uranium 

Facilities, EGG-2530. [SRDB Ref ID: 1781]

Healy, J. W. Surface Contamination: Decision Levels. LA-4558-MS. Los Alamos, New Mexico: Los 

Alamos Scientific Laboratory; 1971.

Leavitt, Michael O., “HHS Designation of Additional Members of the Special Exposure Cohort under 

The Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from Allied Chemical Corporation Plant, Metropolis, Illinois,” February 1, 

2007.

Linsley, G. S. Resuspension of the Transuranium Elements – A review of Existing Data; NRPB-R75. 

Harwell, Didcot, Oxon: 1978.

NCRP (National Council on Radiation Protection and Measurements), 1987, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland.

NIOSH (National Institute for Occupational Safety and Health), 2006a, SEC Petition Evaluation 

Report, Petition SEC-00067, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, 

Ohio, November 10.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 30 of 34

NIOSH (National Institute for Occupational Safety and Health), 2006b, Report on Residual 

Radioactive and Beryllium Contamination at Atomic Weapons Employer Facilities and 

Beryllium Vendor Facilities, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

December.

NIOSH (National Institute for Occupational Safety and Health), (2006c) External Dose Reconstruction 

Implementation Guideline, Rev 2, OCAS-IG-001, National Institute for Occupational Safety 

and Health, Office of Compensation Analysis and Support, Cincinnati, Ohio. [SRDB Ref ID: 

29929]

NRC (U. S. Nuclear Regulatory Commission), 1993, "Materials License," License Number SUB-526, 

Amendment 15, for Allied-Signal, with supporting material, August 26. [SRDB Ref ID: 12377, 

p. 9]

NRC (U. S. Nuclear Regulatory Commission), 2003, "Materials License," License Number SUB-526, 

Amendment No. 15, with excerpts from Honeywell Application for Renewal of Source Materials 

License SUB-526, Docket 40-3392, January 30. [SRDB Ref ID: 12380]

ORAUT (Oak Ridge Associated Universities Team), 2005a, An Exposure Matrix for Bridgeport Brass: 

Havens Laboratory and Adrian Plant, ORAUT-TKBS-0030, Rev. 00, Oak Ridge, Tennessee, 

September 15. [SRDB Ref ID: 19476] 

ORAUT (Oak Ridge Associated Universities Team), 2005b, Estimation of Neutron Dose Rates from 

Alpha-Neutron Reactions in Uranium and Thorium Compounds, ORAUT-OTIB-0024, Rev. 00, 

Oak Ridge, Tennessee, April 7. [SRDB Ref ID: 19445] 

ORAUT (Oak Ridge Associated Universities Team), 2005c, Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge, 

Tennessee, December 21. [SRDB Ref ID: 20220] 

ORAUT (Oak Ridge Associated Universities Team), 2006a, A Standard Complex-Wide Correction 

Methodology for Overestimating External Doses Measured with Film Badge Dosimeters, 

ORAUT-OTIB-0010, Rev. 01, Oak Ridge, Tennessee, June 5. [SRDB Ref ID: 29953] 

ORAUT (Oak Ridge Associated Universities Team), 2006b, External Dose Reconstruction, ORAUTPROC-0006, Rev. 01, Oak Ridge, Tennessee, June 5. [SRDB Ref ID: 29985] 

ORAUT (Oak Ridge Associated Universities Team), 2006c, Estimating the Maximum Plausible Dose 

to Workers at Atomic Weapons Employer Facilities, ORAUT-OTIB-0004, Rev. 03 PC-2, Oak 

Ridge, Tennessee, December 6. [SRDB Ref ID: 29949] 

ORAUT (Oak Ridge Associated Universities Team), 2006d, Portsmouth Gaseous Diffusion Plant – 

Occupational External Dose, ORAUT-TKBS-0015-6, Rev. 01, Oak Ridge, Tennessee, 

November 7. [SRDB Ref ID: 30055] 

Perkins, B. L., 1982, Evaluation of Environmental Control Technologies for Commercial Nuclear Fuel 

Conversion (UF6) Facilities, LA-9397-MS, University of California, Los Alamos National 

Laboratory, Los Alamos, New Mexico, October. [SRDB Ref ID: 12379]
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RFETS (Rocky Flats Environmental Technology Site), 2000, Report on the Flow of Recycled Uranium 

at the Department of Energy's Rocky Flats Plant, 1953 - 1993, Golden, Colorado, June 30. 

[SRDB Ref ID: 16227]

Rogers, R., 2006, “Honeywell Metropolis Bioassay Report”, November 10. [SRDB Ref ID: 32707] 

Sehmel, G. A. Particle Resuspension: A Review. Environmental International 4:107-127; 1980.

Sloop, J. L., 1978, "Liquid Hydrogen as a Propulsion Fuel, 1945-1959," SP-4404, National 

Aeronautics and Space Administration, Washington, D.C., 

http://www.hq.nasa.gov/office/pao/History/SP-4404/ch5-6.htm, accessed December 29, 2005. 

[SRDB Ref ID: 24144]

Thomas, E. M., 2007, e-mail to C. Bloom, "Project Guidance for AWE Chest X-Rays," ORAU Team, 

Cincinnati, Ohio, January 30. 

Till, J. E.; Meyer, H. R. Radiological Assessment. A Textbook on Environmental Dose Analysis. 

NUREG/CR-3332. NUREG/CR-3332. Washington, DC: U.S. Nuclear Regulatory Commission; 

1983.

Wilkins, C. W., 1992, Semiannual Health Physics Data Review (July 1991 - December 1991), March 

30. [SRDB Ref ID: 12374, p. 58]

Wilson, R. H., 1958, “The Hanford Uranium Bio-assay Program,” Symposium on Occupational Health 

Experience and Practices in the Uranium Industry, Held in New York City, October 15 – 17, 

1958, HASL-58, U.S. Atomic Energy Commission, New York Operations Office, New York, 

New York, September, p. 77. [SRDB Ref ID: 7886, p. 89] 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 32 of 34

ATTACHMENT A

APPLICATION OF RESIDUAL DOSE FACTORS

Page 1 of 3 

For the purpose of demonstration of the application of residual dose factors two example cases are 

provided. In all cases, the internal dose is first assessed using the worker’s urine and lung count data 

(if available) during the entire employment period, or up to date of diagnosis. Once an appropriate set 

of intake regimes are determined, adjustments are made as follows:

- Break the intake regimes into the following time periods: 

o pre 1977

o 1977

o 1978

o 1979 onward. 

- Adjust post 1976 uranium intake rates using appropriate factor in Table 5-2 (based on intake 

year)

- Add intakes for uranium progeny based on ratios in Table 5-1 for post 1976 periods

Example 1.

A chronic uranium intake of 100 pCi/d is to be assigned from June 1, 1972 through March 25, 1983. 

This example is typical for assignment of missed dose only.

- Break this single chronic intake into appropriate time periods:

o June 1, 1972 – December 31, 1976 100 pCi/d (Uranium)

o January 1, 1977 – December 31, 1977 100 pCi/d (Uranium)

o January 1, 1978 – December 31, 1978 100 pCi/d (Uranium)

o January 1, 1979 – March 25, 1983 100 pCi/d (Uranium)

- Adjust the post 1976 uranium intake rates based on factors in Table 5-2 

o June 1, 1972 – December 31, 1976 100 pCi/d (Uranium)

o January 1, 1977 – December 31, 1977 100 pCi/d (Uranium)

o January 1, 1978 – December 31, 1978 3 pCi/d (Uranium)

o January 1, 1979 – March 25, 1983 0.07 pCi/d (Uranium)

- Add intakes for uranium progeny for post 1976 intake regimes using appropriate ratios in 

Table 5-1. 
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 33 of 34

ATTACHMENT A

APPLICATION OF RESIDUAL DOSE FACTORS

Page 2 of 3 

Intake period Radionuclide Intake rate, pCi/day

June 1, 1972 – December 31, 1976 U-234 100

January 1, 1977 – December 31, 1977

U-234 100

Th-232 7.6

Th-230 16.7

Th-228 7.6

Ra-226 0.846

Pb-210 0.846

Po-210 0.846

January 1, 1978 – December 31, 1978

U-234 3

Th-232 0.23

Th-230 0.5

Th-228 0.23

Ra-226 0.025

Pb-210 0.025

Po-210 0.025

January 1, 1979 – March 25, 1983

U-234 0.07

Th-232 0.0053

Th-230 0.0117

Th-228 0.0053

Ra-226 0.0006

Pb-210 0.0006

Po-210 0.0006

Example 2. 

Acute uranium intakes are to be assigned as follows: 10,000 pCi on March 5, 1973; 100,000 pCi on 

January 15, 1977; and 100,000 pCi on February 15, 1978. Uranium intakes are assigned:

o March 15, 1973 10,000 pCi (Uranium)

o January 15, 1977 100,000 pCi (Uranium)

o February 15, 1978 100,000 pCi (Uranium)

- Adjust the post 1976 uranium intake rates based on factors in Table 5-2 

o March 15, 1973 10,000 pCi (Uranium)

o January 15, 1977 100,000 pCi (Uranium)

o February 15, 1978 3,000 pCi (Uranium)

- Add intakes for uranium progeny for post 1976 intake regimes using appropriate ratios in 

Table 5-1.
Document No. ORAUT-TKBS-0044 Revision No. 01 Effective Date: 10/01/2007 Page 34 of 34

ATTACHMENT A

APPLICATION OF RESIDUAL DOSE FACTORS

Page 3 of 3 

Intake date Radionuclide Intake, pCi

March 15, 1973 U-234 10,000

January 15, 1977

U-234 100,000

Th-232 7,600

Th-230 16,700

Th-228 7,600

Ra-226 846

Pb-210 846

Po-210 846

February 15, 1978

U-234 3,000

Th-232 228

Th-230 501

Th-228 228

Ra-226 25

Pb-210 25

Po-210 25

Note that for this example, it is also necessary to calculate missed dose (as in Example 1) and 

perform a missed versus fitted comparison, in accordance with section 5.5.3.1 of ORAUT-OTIB-0060. 

This comparison should be performed based on the adjusted intakes (considering Table 5-1 and 5-2 

adjustments, as appropriate).